Eung Soo Kim

Design Option of Heat Exchanger for the Next Generation Nuclear Plant

The next generation nuclear plant (NGNP), a very high temperature gas-cooled reactor (VHTR) concept, will provide the first demonstration of a closed-loop Brayton cycle at a commercial scale, producing a few hundred megawatts of power in the form of electricity and hydrogen. The power conversion unit for the NGNP will take advantage of the significantly higher reactor outlet temperatures of the VHTRs to provide higher efficiencies than can be achieved with the current generation of light water reactors. Besides demonstrating a system design that can be used directly for subsequent commercial deployment, the NGNP will demonstrate key technology elements that can be used in subsequent advanced power conversion systems for other Generation IV reactors. In anticipation of the design, development, and procurement of an advanced power conversion system for the NGNP, the system integration of the NGNP and hydrogen plant was initiated to identify the important design and technology options that must be considered in evaluating the performance of the proposed NGNP. As part of the system integration of the VHTRs and the hydrogen production plant, the intermediate heat exchanger is used to transfer the process heat from VHTRs to the hydrogen plant. Therefore, the design and configuration of the intermediate heat exchanger are very important. This paper describes analyses of one stage versus two-stage heat exchanger design configurations and simple stress analyses of a printed circuit heat exchanger (PCHE), helical-coil heat exchanger, and shell-and-tube heat exchanger.

Process Heat Exchanger Options for the Advanced High Temperature Reactor

The work reported herein is a significant intermediate step in reaching the final goal of commercial-scale deployment and usage of molten salt as the heat transport medium for process heat applications. The primary purpose of this study is to aid in the development and selection of the required heat exchanger for power production and process heat application, which would support large-scale deployment.

Process Heat Exchanger Options for Fluoride Salt High Temperature Reactor

The work reported herein is a significant intermediate step in reaching the final goal of commercial-scale deployment and usage of molten salt as the heat transport medium for process heat applications. The primary purpose of this study is to aid in the development and selection of the required heat exchanger for power production and process heat application, which would support large-scale deployment.

High Temperature Thermal Devices for Nuclear Process Heat Transfer Applications

1.1 Heat exchangers A heat exchanger is a component used to transfer heat from one medium to another. The media may be separated by a solid wall, so they never mix, or they may be in direct contact (Kakac and Liu 2002). Heat exchangers generally have no external heat nor work interactions and are typically used in the following applications: • Space heating • Refrigeration • Air conditioning • Power plants • Chemical plants • Petrochemical plants • Petroleum refineries • Natural gas processing.

Validations of CFD Code for Density-Gradient Driven Air Ingress Stratified Flow

Air ingress into a very high temperature gas-cooled reactor (VHTR) is an important phenomenon to consider because the air oxidizes the reactor core and lower plenum where the graphite structure supports the core region in the gas turbine modular helium reactor (GT-MHR) design, thus jeopardizing the reactor’s safety. Validating the computational fluid dynamics (CFD) code used to analyze the air ingress phenomena is therefore an essential part of the safety analysis and the ultimate computation required for licensing. An experimental data set collected by ETH Zurich on a lock exchange experiment (Grobelbauser et al., Lowe et al. 2002; Lowe et al. 2005; and Shin et al. 2004) was selected for the validation. The experiment was based on a series of lock exchange flows with gases of different density ratios varying from 0.046 to 0.9 in a closed channel of a square cross-section. The focus was on the quantitative measurement of front velocities of the gravity current flows. The experiment results cover the full range of gas intrusions—heavy as well as light—for the gravity current flows in the lock exchange situations. FLUENT CFD code (ANSYS Fluent 2008) was used. The calculated results showed very good agreement with the experimental data. A number of tables and comparison plots are included to summarize the estimated current speeds. The current speed obtained by experimental data was 1.25 m/s and that of the simulation was 1.19 m/s. This result indicates that the deviation of the simulation is only 4.8% that of the experimental data.Copyright

PIV Uncertainty Methodologies for CFD Code Validation at the MIR Facility

Currently, computational fluid dynamics (CFD) is widely used in the nuclear thermal hydraulics field for design and safety analyses. To validate CFD codes, high quality multi dimensional flow field data are essential. The Matched Index of Refraction (MIR) Flow Facility at Idaho National Laboratory has a unique capability to contribute to the development of validated CFD codes through the use of Particle Image Velocimetry (PIV). The significance of the MIR facility is that it permits non intrusive velocity measurement techniques, such as PIV, through complex models without requiring probes and other instrumentation that disturb the flow. At the heart of any PIV calculation is the cross-correlation, which is used to estimate the displacement of particles in some small part of the image over the time span between two images. This image displacement is indicated by the location of the largest peak. In the MIR facility, uncertainty quantification is a challenging task due to the use of optical measurement techniques. Currently, this study is developing a reliable method to analyze uncertainty and sensitivity of the measured data and develop a computer code to automatically analyze the uncertainty/sensitivity of the measured data. The main objective of this study is to develop a well established uncertainty quantification method for the MIR Flow Facility, which consists of many complicated uncertainty factors. In this study, the uncertainty sources are resolved in depth by categorizing them into uncertainties from the MIR flow loop and PIV system (including particle motion, image distortion, and data processing). Then, each uncertainty source is mathematically modeled or adequately defined. Finally, this study will provide a method and procedure to quantify the experimental uncertainty in the MIR Flow Facility with sample test results.

Small Modular Molten Salt Reactor (SM-MSR)

The strategic goal of the Small Modular Molten Salt Reactor (SM-MSR) is to broaden the environmental and economic benefits of nuclear energy in the United States by producing power to meet growing energy demands and demonstrating its applicability to market sectors not being served by light water reactors. Of primary importance is demonstrating that the SM-MSR can be operated safely and compete economically with larger reactors. Reactor outlet temperatures (ROTs) of SM-MSRs will likely be much higher (around 700°C) than light water reactors, thereby increasing the efficiency of electricity production and potentially providing process heat for industrial applications, which will help offset the higher per kilowatt costs generally associated with smaller reactors, making the SM-MSR more economically competitive. This paper compares thermal power cycles for given ROT, compares thermal performance using figure of merits and sensitivity study and discusses the comparative advantages of SM-MSRs.Copyright

Air Ingress Analysis: Computational Fluid Dynamics Models

Idaho National Laboratory (INL), under the auspices of the U.S. Department of Energy (DOE), is performing research and development that focuses on key phenomena important during potential scenarios that may occur in very high temperature reactors (VHTRs). Phenomena identification and ranking studies to date have ranked an air ingress event, following on the heels of a VHTR depressurization, as important with regard to core safety. Consequently, the development of advanced air-ingress-related models and verification and validation data are a very high priority. Following a loss of coolant and system depressurization incident, air will enter the core of the High Temperature Gas Cooled Reactor through the break, possibly causing oxidation of the core and reflector graphite structure. Simple core and plant models indicate that, under certain circumstances, the oxidation may proceed at an elevated rate with additional heat generated from the oxidation reaction itself. Under postulated conditions of fluid flow and temperature, excessive degradation of lower plenum graphite can lead to a loss of structural support. Excessive oxidation of core graphite can also lead to a release of fission products into the confinement, which could be detrimental to reactor safety. Computational fluid dynamics models developed in this study will improve our understanding of this phenomenon. This paper presents two-dimensional (2-D) and three-dimensional (3-D) computational fluid dynamic (CFD) results for the quantitative assessment of the air ingress phenomena. A portion of the results from density-driven stratified flow in the inlet pipe will be compared with the experimental results.Copyright

Computational Fluid Dynamics Analyses on Very High Temperature Reactor Air Ingress

A preliminary computational fluid dynamics (CFD) analysis was performed to understand density-gradient-induced stratified flow in a Very High Temperature Reactor (VHTR) air-ingress accident. Various parameters were taken into consideration, including turbulence model, core temperature, initial air mole-fraction, and flow resistance in the core. The gas turbine modular helium reactor (GT-MHR) 600 MWt was selected as the reference reactor and it was simplified to be 2-D geometry in modeling. The core and the lower plenum were assumed to be porous bodies. Following the preliminary CFD results, the analysis of the air-ingress accident has been performed by two different codes: GAMMA code (system analysis code, Oh et al. 2006) and FLUENT CFD code (Fluent 2007). Eventually, the analysis results showed that the actual onset time of natural convection (~160 sec) would be significantly earlier than the previous predictions (~150 hours) calculated based on the molecular diffusion air-ingress mechanism. This leads to the conclusion that the consequences of this accident will be much more serious than previously expected.

Stress Analyses of Intermediate Heat Transfer Loop

The Department of Energy and the Idaho National Laboratory are developing a Next Generation Nuclear Plant (NGNP) to serve as a demonstration of state-of-the-art nuclear technology. The purpose of the demonstration is two fold 1) efficient low cost energy generation and 2) hydrogen production. Although a next generation plant could be developed as a single-purpose facility dedicated to hydrogen production, early designs are expected to be dual purpose. While hydrogen production and advanced energy cycles are still in its early stages of development, research towards coupling a high temperature reactor with electrical generation and hydrogen production is under way. Many aspects of the NGNP must be researched and developed in order to make recommendations on the final design of the plant. Parameters such as working conditions, cycle components, working fluids, and power conversion unit configurations must be understood. A number of configurations of the power conversion unit were demonstrated in this study. An intermediate heat transport loop for transporting process heat to a High Temperature Steam Electrolysis (HTSE) hydrogen production plant was used. Helium, CO2 , and a 80% nitrogen, 20% helium mixture (by weight) were studied to determine the best working fluid in terms cycle efficiency and development cost. In each of these configurations the relative component sizes were estimated for the different working fluids. Parametric studies were carried out on reactor outlet temperature, mass flow, pressure, and turbine cooling. Recommendations on the optimal working fluid for each configuration were made. Engineering analyses were performed for several configurations of the intermediate heat transport loop that transfers heat from the nuclear reactor to the hydrogen production plant. The analyses evaluated parallel and concentric piping arrangements and two different working fluids, including helium and a liquid salt. The thermal-hydraulic analyses determined the size and insulation requirements for the hot and cold leg pipes in the different configurations. Mechanical analyses were performed to determine hoop stresses and thermal expansion characteristics for the different configurations.Copyright